Multiphysics are simulations using computer applications or software to couple multiple physical events in order to predict or validate the physical real-world outcome.
The nuclear industry has always prioritised the safe, reliable and economically attractive operation of nuclear power reactors. Given these priorities, the development, validation and application of predictive reliable modelling capabilities for both normal and accident conditions has evolved from best-estimate calculations to first principle high-fidelity multi-physics simulations.
Especially important are the multi-physics interactions in reactor cores. In the past, these different interactions were treated either as boundary conditions (i.e. each physics calculation was performed independently and the impact of other physics phenomena were taken into account through boundary conditions) or using very simplistic models for some of the physics phenomena. Examples for the latter are the point kinetics model implemented in system thermal-hydraulic models or one-dimensional thermal-hydraulic models implemented in neutronics core design codes.
Recently, advances in computing power and numerical methods have broadened the appeal of multi-physics simulations, in turn highlighting the need for experiments to validate these simulations.
A community of multi-physics experts was formed within the Expert Group on System Reactor Multi-physics (EGMUP), and is responsible for advancing multi-physics activities.
Under the EGMUP, the NEA has undertaken activities aimed at verification and validation of multi-physics tools, and collecting the experimental data that underpins the work.
Notably, the EGMUP publishes expert guidance in the domain of multiphysics, and work conducted within EGMUP has led to numerous conference and journal publications.
The Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of Light Water Reactors (LWRs) is an international high-visibility benchmark for uncertainty analysis in best-estimate coupled code calculations for design, operation, and safety analysis of LWRs. The annual workshops are attended by many experts in industry, research institutes, national laboratories, academia, and government agencies.
The goal of the Benchmark for Uncertainty Analysis in Best-Estimate Modelling for Design, Operation and Safety Analysis of Light Water Reactors (LWR-UAM) was to determine the uncertainty in light water reactor (LWR) systems and processes in all stages of calculations. It was estimated through a simulation process of nine exercises in three phases provided by the benchmarking framework.
Helium-cooled very high-temperature gas reactors are highlighted as a key technology with the potential to improve the competitiveness of nuclear energy within the Generation IV International Forum (GIF). Developing tools and methods to support this technology is seen as a priority by NEA member countries.
The expert group deals with the activities associated with the certification of experimental data and benchmark models along with establishing the processes and procedures for using the data and benchmark models for validation of modelling and simulation tools and data.
The expert group provides advice to the Working Party on Scientific Issues of Reactor Systems (WPRS) and the nuclear community on the scientific development needs (data and methods, validation experiments, scenario studies) of sensitivity and uncertainty methodology for modelling of different reactor systems and scenarios.
This benchmark was a continuation of the V1000CT activities and defined a coupled code problem for further validation of thermal-hydraulics system codes for application with Russian-designed VVER-1000 reactors based on actual plant data from the Russian nuclear power plant Kalinin Unit 3 (Kalinin-3)
This benchmark incorporated full 3-D modelling of the reactor core into system transient codes for best-estimate simulations of the interactions between reactor core behaviour and plant dynamics and their testing on a number of transients of importance for plant behaviour and safety analysis.
A number of tests with detail well documented neutronics and thermal-hydraulics measurements data have been performed at the Rostov Unit 2 (Rostov-2) nuclear power plant (NPP). The reactor type is a VVER-1000 with fuel assemblies of type TBC-2M, which enable an 18-month fuel cycle length.
A sodium-cooled fast reactor (SFR) is a fast neutron reactor that uses molten sodium metal as a coolant. Sodium-cooled Fast Reactors (SFRs) are the most promising type of reactors to achieve Generation IV (Gen IV) nuclear reactor goals at a reasonable time scale given the accumulated experience over the years.
The Subgroup on Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFR-UAM) was formed to check the use of best-estimate codes and data.